Material Properties of Reactor Pressure Vessel Shells Affected by Hydrogen Flaking

2016 ◽  
Author(s):  
Robert Gérard ◽  
Michel De Smet ◽  
Rachid Chaouadi
Keyword(s):  
Fracture Toughness ◽  
Material Properties ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Operating Cycle ◽  
Safety Case ◽  
Safety Cases ◽  
Definition Of ◽  
Reactor Pressure ◽  
Trend Curve

During the summer outages of 2012, large numbers of nearly-laminar indications were found in the core shells of the Doel 3 and Tihange 2 reactor pressure vessels (RPV). As a consequence, both units remained in cold shutdown with their core unloaded. A series of examinations, tests and inspections were performed leading to the conclusion that the indications are hydrogen flakes and that they do not affect the structural integrity of the RPV, regardless of the operating mode, transient or accident condition. All this was documented in the Safety Case reports issued in December 2012 and in the Safety Case Addenda issued in April 2013 [1]. Based on those reports, the Belgian Federal Agency for Nuclear Control (FANC) authorized the restart of both units which went back on-line in June 2013. A key input required for this Safety Case was the definition of the appropriate material properties, in particular fracture toughness, for the RPV shells affected by hydrogen flakes. A material testing program on non-irradiated materials evaluated aspects like the possible effects of macro-segregations and local segregations (ghost lines) and of specimen orientation on the fracture toughness. The irradiation embrittlement sensitivity of the zone of macro-segregation in which the flakes are located was evaluated on the basis of the maximum enrichment in Cu, P and Ni in macro-segregations based on literature data. This was the basis of the trend curve of RTNDT evolution vs. fluence used in the Safety Cases submitted in 2012–2013. The restart authorization in 2013 was accompanied by a number of “mid-term” requirements, to be completed during the first operating cycle after the restart. One of these requirements was the mechanical testing of irradiated specimens containing hydrogen flakes, in order to confirm the conservativeness of the RTNDT trend curve used for the structural integrity analyses. After a first irradiation campaign of a material containing hydrogen flakes in the BR2 reactor of the Belgian Nuclear Research Center SCK.CEN, atypical results were obtained and the utility decided to shut down the units in March 2014. Detailed investigations involving three additional irradiation campaigns in BR2 including other reference materials, among which another material affected by hydrogen flakes, were performed in order to characterize this atypical behaviour and to derive a new conservative RTNDT trend curve. The resulting trend curve was accepted by the FANC and was used in the 2015 Safety Cases [1]. An overview of the Doel 3 and Tihange 2 safety cases is given in [6]. The paper summarizes the results of the material investigations on non-irradiated and irradiated materials and the process leading to the definition of this conservative RTNDT trend curve.

Overview of the Doel 3 and Tihange 2 Reactor Pressure Vessel Safety Cases

2016 ◽  
Author(s):  
Michel De Smet ◽  
Jean Van Vyve
Keyword(s):  
Structural Integrity ◽  
Case Reports ◽  
Operating Mode ◽  
Pressure Vessels ◽  
Irradiation Embrittlement ◽  
Safety Case ◽  
On Line ◽  
Safety Cases ◽  
Integrity Analysis ◽  
Reactor Pressure

During the summer outages of 2012, large numbers of indications were found inside the shell material of the Doel 3 and Tihange 2 reactor pressure vessels (RPV). Therefore both plants remained in cold shutdown with their core unloaded. A series of examinations, tests and inspections were performed leading to the conclusion that the indications are hydrogen flakes and that they do not affect the structural integrity of the RPVs, regardless the operating mode, transient or accident condition. All this was documented in the Safety Case Reports issued in December 2012 and the Safety Case Addenda issued in April 2013. Based on those reports, the Belgian Federal Agency for Nuclear Control (FANC) authorized the restart of both units that went back on-line in June 2013. In parallel, a number of requirements from the FANC were addressed such as the qualification of the applied Ultrasonic Testing (UT) procedure, and mechanical testing of irradiated specimens containing hydrogen flakes. The preliminary tests showed unexpected results regarding the shift in RTNDT (Reference Temperature for Nil Ductility Transition) under irradiation that could not confirm the hypotheses considered in the initial Safety Case Reports. Therefore, the Licensee Electrabel decided to shut down both plants immediately. In order to fully address this concern, the material test programme was extended including several RPV materials and covering additional irradiation campaigns. This led to a modification of the irradiation embrittlement trend curves considered in the structural integrity analysis. In addition, the qualification of the UT procedure led to an updated cartography of the flakes. The structural integrity assessments of both RPVs were revised accordingly. The final Safety Case Reports, confirming the fitness-for-continued operation of both RPVs, were submitted to the FANC in October 2015. FANC allowed restart of both units on November 17th, 2015. The paper gives a historical overview of the Doel 3 and Tihange 2 RPV Safety Cases and explains how the roadmap was built in order to demonstrate the RPV’s structural integrity in the presence of hydrogen flakes.

ORNL Evaluation of Safety Cases for Two Belgian Reactor Pressure Vessels Containing Quasi-Laminar Defects

2017 ◽  
Author(s):  
Hilda B. Klasky ◽  
B. Richard Bass ◽  
Terry L. Dickson ◽  
Sarma B. Gorti ◽  
Randy K. Nanstad ◽  
...  
Keyword(s):  
New York ◽  
Finite Element ◽  
Structural Integrity ◽  
National Laboratory ◽  
Relevant Literature ◽  
Pressure Vessels ◽  
Oak Ridge ◽  
Safety Margins ◽  
Safety Cases ◽  
Reactor Pressure

The Oak Ridge National Laboratory (ORNL) performed a detailed technical review of the 2015 Electrabel (EBL) Safety Cases prepared for the Belgium reactor pressure vessels (RPVs) at Doel 3 and Tihange 2 (D3/T2). The Federal Agency for Nuclear Control (FANC) in Belgium commissioned ORNL to provide a thorough assessment of the existing safety margins against cracking of the RPVs due to the presence of almost laminar flaws found in each RPV. Initial efforts focused on surveying relevant literature that provided necessary background knowledge on the issues related to the quasi-laminar flaws observed in D3/T2 reactors. Next, ORNL proceeded to develop an independent quantitative assessment of the entire flaw population in the two Belgian reactors according to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, Appendix G, “Fracture Toughness Criteria for Protection Against Failure,” New York (both 1992 and 2004 versions). That screening assessment of the EBL-characterized flaws in D3/T2 used ORNL tools, methodologies, and the ASME Code Case N-848, “Alternative Characterization Rules for Quasi-Laminar Flaws”. Results and conclusions derived from comparisons of the ORNL flaw acceptance assessments of D3/T2 with those from the 2015 EBL Safety Cases are presented in the paper. The ORNL screening analyses identified fewer flaws than EBL that were not compliant with the ASME Section XI (1992) criterion; the EBL criterion imposed additional conservatisms not included in ASME Section XI. Furthermore, ORNL’s application of the updated ASME Section XI (2004) criterion produced only four non-compliant flaws, all due to design-basis loss-of-coolant loading transients. Among the latter, only one flaw remained non-compliant when analyzed using the warm-prestress (WPS) cleavage fracture model typically applied in USA flaw assessments. ORNL’s independent refined analysis of that flaw (#1660, which was also non-compliant in the EBL screening assessments) rendered it compliant when modeled as a more realistic individual quasi-laminar flaw using a 3-dimensional XFEM (eXtended Finite Element Method) approach available in the ABAQUS© finite element code. Taken as a whole, the ORNL-specific results and conclusions confirmed the structural integrity of Doel 3 and Tihange 2 under all design transients with ample margin in the presence of the 16,196 detected flaws.

Structural Integrity Assessment of Doel 3 and Tihange 2 RPVs Affected by Hydrogen Flakes: Refined X-FEM Analyses

2016 ◽  
Author(s):  
Pierre Dulieu ◽  
Valéry Lacroix
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Extended Finite Element Method ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Extended Finite Element ◽  
Integrity Assessment ◽  
Safety Case ◽  
Component Manufacturing ◽  
Reactor Pressure

During the 2012 outage at Doel 3 and Tihange 2 Nuclear Power Plants, specific ultrasonic in-service inspections revealed a large number of quasi-laminar indications in the base metal of the reactor pressure vessels, mainly in the lower and upper core shells. The observed indications could subsequently be attributed to hydrogen flaking induced during the component manufacturing process. As a consequence, a Flaw Acceptability Assessment had to be performed as a part of the Safety Case demonstrating the fitness-for-service of these units. In that framework, detailed analyses using eXtended Finite Element Method were conducted to model the specific character of hydrogen flakes. Their quasi-laminar orientation as well as their high density required setting up 3D multi-flaws model accounting for flaw interaction. These calculations highlighted that even the most penalizing flaw configurations are harmless in terms of structural integrity despite the consideration of higher degradation of irradiated material toughness.

Development of Probabilistic Evaluation Models of Fracture Toughness KIc and KIa for Japanese RPV Steels

2015 ◽  
Author(s):  
Jinya Katsuyama ◽  
Genshichiro Katsumata ◽  
Kunio Onizawa ◽  
Kazuya Osakabe ◽  
Kentaro Yoshimoto
Keyword(s):  
Fracture Toughness ◽  
Structural Integrity ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Confidence Limits ◽  
Lognormal Distributions ◽  
Integrity Assessment ◽  
Probabilistic Evaluation ◽  
Evaluation Models ◽  
Reactor Pressure

Probabilistic fracture mechanics (PFM) analysis code PASCAL3 has been developed to apply the PFM analysis to the structural integrity assessment of domestic reactor pressure vessels (RPVs). In this paper, probabilistic evaluation models of fracture toughness KIc and KIa which have the largest scatter among the associated factors based on the database of Japanese RPV steels are presented. We developed probabilistic evaluation models for KIc and KIa based on the Weibull and lognormal distributions, respectively. The models are compared with the existing lower bound of fracture toughness in the Japanese code and probabilistic model in USA. As the results, the 5% confidence limits of the models established in present work corresponded to lower bounds of fracture toughness in the Japanese code. The comparison in the models between present work and USA showed significant differences that may have an influence on fracture probability of RPV.

Study on Application of PFM Analysis Method to Japanese Code for RPV Integrity Assessment Under PTS Events

2015 ◽  
Author(s):  
Kazuya Osakabe ◽  
Koichi Masaki ◽  
Jinya Katsuyama ◽  
Genshichiro Katsumata ◽  
Kunio Onizawa ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Input Data ◽  
Structural Integrity ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Analysis Method ◽  
Integrity Assessment ◽  
Analysis Models ◽  
Reactor Pressure ◽  
The U.S

A probabilistic fracture mechanics (PFM) analysis method for pressure boundary components is useful to evaluate the structural integrity in a quantitative way. This is because the uncertainties related to influence parameters can be rationally incorporated in PFM analysis. From this viewpoint, the probabilistic approach evaluating through-wall cracking frequencies (TWCFs) of reactor pressure vessels (RPVs) has already been adopted as the regulation on fracture toughness requirements against PTS events in the U.S. As a study of applying PFM analysis to the integrity assessment of domestic RPVs, JAEA has been preparing input data and analysis models to calculate TWCFs using PFM analysis code PASCAL3. In this paper, activities have been introduced such as preparing input data and models for domestic RPVs, verification of PASCAL3, and formulating guideline on general procedures of PFM analysis for the purpose of utilizing PASCAL3. In addition, TWCFs for a model RPV evaluated by PASCAL3 are presented.

Probabilistic Structural Integrity Evaluation on a Pressurized Water Reactor Pressure Vessel Under Pressure–Temperature Limit Operations

2015 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Temperature Limit ◽  
Pressure Vessels ◽  
Pressurized Water Reactor ◽  
Operational Flexibility ◽  
Pressurized Water ◽  
Water Reactor ◽  
Reactor Pressure

The normal reactor startup (heat-up) and shut-down (cool-down) operation limits are defined by the ASME Code Section XI-Appendix G, to ensure the structural integrity of the embrittled nuclear reactor pressure vessels (RPVs). In the paper, the failure risks of a Taiwan domestic pressurized water reactor (PWR) pressure vessel under various pressure-temperature limit operations are analyzed. Three types of pressure-temperature limit curves established by different methodologies, which are the current operation limits of the domestic RPV based on the KIa fracture toughness curve in 1998 or earlier editions of ASME Section XI-Appendix G, the recently proposed limits according to the KIC fracture toughness curve after the 2001 edition of ASME Section XI-Appendix G, and the risk-informed revision method proposed in MRP-250 report that provides more operational flexibility, are considered. The ORNL’s probabilistic fracture mechanics code, FAVOR, is employed to perform a series of fracture probability analyses for the RPV at multiple levels of embrittlement under such pressure-temperature limit transients. The analysis results indicate that the pressure-temperature operation limits associated with more operational flexibility will result in higher failure risks to the RPV. The shallow inner surface breaking flaw due to the clad fabrication defect is the most critical factor and dominates the failure risk of the RPV under pressure-temperature limit operations. Present work can provide a risk-informed reference for the safe operation and regulation of PWRs in Taiwan.

Neutron Embrittlement Aging Management of Nuclear Reactor Pressure Vessels

2003 ◽  
Author(s):  
William L. Server
Keyword(s):  
Fracture Toughness ◽  
Transition Temperature ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Temperature Shift ◽  
Environmental Parameters ◽  
Pressure Vessels ◽  
Reactor Vessel ◽  
Vessel Material ◽  
Reactor Pressure

The management of neutron embrittlement of nuclear reactor pressure vessels involves monitoring of the changes in the fracture toughness of surveillance capsule specimens that closely approximate the actual reactor vessel material(s). The measurement of fracture toughness is currently performed in an indirect manner using Charpy V-notch impact specimens, although the direct measurement of fracture toughness is possible using the same small Charpy specimens fatigue precracked to produce acceptable fracture toughness three-point bend specimens. This paper first examines the current Charpy-based approach and the development of a recent embrittlement correlation that has been incorporated into ASTM E 900-02, “Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials.” This correlation provides the latest mechanistically-guided approach to assess the changes in transition temperature shift. This same correlation and mechanistic guidance can be used with measured fracture toughness data developed following ASTM E 1921-02 to account for differences in surveillance material versus actual vessel material. Additionally, environmental parameters such as fluence and temperature also can be adjusted between different irradiation facilities using this latest correlation. This paper focuses on the application of the new ASTM E 900-02 correlation to Charpy-based and fracture toughness-based measurements to develop the best predictive approach for assuring structural integrity of reactor vessel materials. Key technical issues important for extended vessel life also are discussed.

Applicability of a Crack Length Correction on the RCC-M Fracture Toughness Reference Curve

2016 ◽  
Author(s):  
Isabelle Delvallée-Nunio ◽  
Christophe Blain
Keyword(s):  
Fracture Toughness ◽  
Size Effect ◽  
Crack Length ◽  
Pressure Vessels ◽  
Nuclear Safety ◽  
Radiological Protection ◽  
Reference Curve ◽  
Temperature Dependent ◽  
Safety Cases ◽  
Reactor Pressure

The safety cases of the French 900-MWe and 1300-MWe reactor pressure vessels (RPV) have been reviewed respectively in 2010 and 2015 by the French Nuclear Safety Authority (ASN) and the Institute for Radiological Protection and Nuclear Safety (IRSN) in order to decide of the continued operation for the next ten years after their third decennial outage. The demonstration of the RPV serviceability, based on a deterministic approach, includes an assessment of the risk of crack initiation under normal and accidental conditions. The fracture criterion defined by the French utility, takes into account a size effect on fracture toughness in the ductile-to-brittle transition of ferritic steel by the application of a correction factor to the fracture toughness estimated on the French code RCC-M fracture toughness reference curve. This correction related to the crack length is built on the observation that the failure probability follows a Weibull distribution and on the assumption that the RCC-M KIC curve can be associated to a reference crack length of 100 mm. A first crack length correction has been judged inappropriate by ASN and IRSN in 2010. A new temperature-dependent correction has then been proposed in 2014. Nevertheless, numerous elements presented in this paper have led ASN and IRSN to the conclusion that this new correction should not be applied to the RCC-M KIC curve. In particular the RCC-M KIC curve already integrates intrinsically a size effect on fracture toughness.

Comparative analysis and verification of engineering methods of shallow cracks effect for fracture toughness prediction for reactor pressure vessels

2020 ◽  
pp. 140-165
Author(s):  
V. I. Kostylev ◽  
B. Z. Margolin
Keyword(s):  
Experimental Data ◽  
Fracture Toughness ◽  
Comparative Analysis ◽  
Nuclear Reactor ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Local Methods ◽  
European Method ◽  
Shallow Crack ◽  
Reactor Pressure

The main features of shallow cracks fracture are considered, and a brief analysis of methods allowing to predict the temperature dependence of the fracture toughness KJC (T) for specimens with shallow cracks is given. These methods include DA-method, (JQ)-method, (J-T)-method, “local methods” with its multiparameter probabilistic approach, GP method uses power approach, and also two engineering methods – RMSC (Russian Method for Shallow Crack) and EMSC (European Method for Shallow Crack). On the basis of 13 sets of experimental data for national and foreign steels, a detailed verification and comparative analysis of these two engineering methods were carried out on the materials of the VVER and PWR nuclear reactor vessels considering the effect of shallow cracks.

Structural Integrity Research for Reactor Pressure Vessel Under In-Vessel Retention of a Core Melt

2016 ◽  
Author(s):  
Yongjian Gao ◽  
Yinbiao He ◽  
Ming Cao ◽  
Yuebing Li ◽  
Shiyi Bao ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Material Properties ◽  
Power Plants ◽  
Structural Integrity ◽  
Plastic Material ◽  
Plastic Region ◽  
Reactor Pressure Vessel ◽  
Elastic Plastic ◽  
Elastic Plastic Material ◽  
Reactor Pressure

In-Vessel Retention (IVR) is one of the most important severe accident mitigation strategies of the third generation passive Nuclear Power Plants (NPP). It is intended to demonstrate that in the case of a core melt, the structural integrity of the Reactor Pressure Vessel (RPV) is assured such that there is no leakage of radioactive debris from the RPV. This paper studied the IVR issue using Finite Element Analyses (FEA). Firstly, the tension and creep testing for the SA-508 Gr.3 Cl.1 material in the temperature range of 25°C to 1000°C were performed. Secondly, a FEA model of the RPV lower head was built. Based on the assumption of ideally elastic-plastic material properties derived from the tension testing data, limit analyses were performed under both the thermal and the thermal plus pressure loading conditions where the load bearing capacity was investigated by tracking the propagation of plastic region as a function of pressure increment. Finally, the ideal elastic-plastic material properties incorporating the creep effect are developed from the 100hr isochronous stress-strain curves, limit analyses are carried out as the second step above. The allowable pressures at 0 hr and 100 hr are obtained. This research provides an alternative approach for the structural integrity evaluation for RPV under IVR condition.

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